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Journal Articles

Experiments on gas entrainment phenomena due to free surface vortex induced by flow passing beside stagnation region

Ezure, Toshiki; Ito, Kei; Tanaka, Masaaki; Ohshima, Hiroyuki; Kameyama, Yuri*

Nuclear Engineering and Design, 350, p.90 - 97, 2019/08

This paper reports the results of an experiment on surface vortex-type gas entrainment, which occurs in a shear flow area where flow passes besides the stagnation region. The relationship between the free surface dimple shape and the velocity distribution around the free surface vortex was simultaneously grasped under several horizontal and suction velocity conditions by a combination of visualization and particle image velocimetry measurements. The circulation and the vertical velocity gradient were also evaluated from the velocity distributions at a plane just below the free surface and the middle plane between the free surface and suction nozzle. Quantitative relationships between the circulation, vertical velocity gradient, and gas core length were obtained in time-trends as fundamental data to develop the evaluation method of gas entrainment. Furthermore, it was confirmed that the evaluation method based on a vortex model was an effective way to evaluate gas entrainment.

Journal Articles

A Conceptual design study of pool-type sodium-cooled fast reactor with enhanced anti-seismic capability

Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Uchida, Masato*; Miyagawa, Takayuki*; Eto, Masao*; Suzuno, Tetsuji*; Matoba, Ichiyo*; Endo, Junji*; Watanabe, Osamu*; et al.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.

Journal Articles

Subchannel analysis of thermal-hydraulics in a fuel assembly with inner duct structure of a sodium-cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Journal of Nuclear Engineering and Radiation Science, 5(2), p.021001_1 - 021001_12, 2019/04

In the design study of an advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been considered as one of the measures to enhance safety of the reactor during the core disruptive accident. In this study, thermal-hydraulics in FAIDUS was investigated with the in-house subchannel analysis code named ASFRE. Before the application to FAIDUS, applicability of ASFRE to FAs was confirmed through the numerical simulations for the experiments of simulated FA. Through the comparisons between the numerical results of thermal-hydraulic analyses of FAIDUS and a typical FA without the inner duct, it was indicated that significant asymmetric temperature distribution would not occur in FAIDUS at both high and low flow rate conditions.

Journal Articles

Improvement of steam generator tube failure propagation analysis code LEAP for evaluation of overheating rupture

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki

Journal of Nuclear Science and Technology, 56(2), p.201 - 209, 2019/02

 Percentile:100(Nuclear Science & Technology)

Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was developed to expand application range of an existing computer code. Applicability of the method was demonstrated through the numerical analysis of the experiment on water vapor discharging in liquid sodium.

Journal Articles

Development of unstructured mesh-based numerical method for sodium-water reaction phenomenon

Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki

Nuclear Technology, 205(1-2), p.119 - 127, 2019/01

 Percentile:100(Nuclear Science & Technology)

To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile and location of the Mach disk showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium.

Journal Articles

Multi-dimensional numerical investigation of sodium spray combustion; Benchmark analysis of SNL T3 experiment

Sonehara, Masateru; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Clark, A. J.*; Denman, M. R.*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 5 Pages, 2018/11

no abstracts in English

Journal Articles

Study on gas entrainment from unstable drifting vortexes on liquid surface

Hirakawa, Moe*; Kikuchi, Yuichiro*; Sakai, Takaaki*; Tanaka, Masaaki; Ohshima, Hiroyuki

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

Gas entrainment (GE) from cover gas is one of key issue for Sodium-cooled fast reactors to prevent unexpected effects to core reactivity. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortexes on water experiments which were generated as wake vortexes behind a plate obstacle in the circulating water channel. In this paper, the results of comparison between experiments and analyses were discussed and the gas core lengths from the surface vortexes were evaluated by using the evaluation tool named StreamViewer developed by Japan Atomic Energy Agency.

Journal Articles

Visualization study on droplet-entrainment in a high-speed gas jet into a liquid pool

Sugimoto, Taro*; Saito, Shimpei*; Kaneko, Akiko*; Abe, Yutaka*; Uchibori, Akihiro; Ohshima, Hiroyuki

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 7 Pages, 2018/07

A computational fluid dynamics code for a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors has been developed. In order to provide the data for validation of this code, the visualization experiment on liquid droplet entrainment in the high-pressure air jet submerged in the water pool was carried out. The experiment successfully elucidated the behavior, such as atomization of the relatively large diameter liquid droplet generated from the gas-liquid interface.

Journal Articles

Development of numerical analysis method for core thermal-hydraulics during natural circulation decay heat removal in SFR, 1; Validation of ASFRE code in estimation of radial heat transfer phenomena

Kikuchi, Norihiro; Doda, Norihiro; Hashimoto, Akihiko*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki

Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06

For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.

Journal Articles

Application of unstructured mesh-based numerical method to sodium-water reaction phenomenon analysis code SERAPHIM

Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki

Nippon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00394_1 - 17-00394_6, 2018/03

For assessment of the wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an underexpanded jet experiment. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.

Journal Articles

SNL/JAEA collaboration on sodium fire benchmarking

Clark, A. J.*; Denman, M. R.*; Takata, Takashi; Ohshima, Hiroyuki

SAND2017-12409, 39 Pages, 2017/11

Two sodium spray fire experiments performed by Sandia National Laboratories (SNL) were used for a code-to-code comparison between CONTAIN-LMR and SPHINCS. Both computer codes are used for modeling sodium accidents in sodium fast reactors. The comparison between the two codes provides insights into the ability of both codes to model sodium spray fires. The SNL T3 and T4 experiments are 20 kg sodium spray fires with sodium spray temperatures of 200$$^{circ}$$C and 500$$^{circ}$$C, respectively. The vessel in the SNL T4 experiment experienced a rapid pressurization that caused of the instrumentation ports to fail during the sodium spray. Despite these unforeseen difficulties, both codes were shown in good agreement with the experiments. SPHINCS showed better long-term agreement with the SNL T3 experiment than CONTAIN-LMR. The unexpected port failure during the SNL T4 experiment presented modelling challenges.

Journal Articles

Development of unstructured mesh-based numerical method for sodium-water reaction phenomenon in steam generators of sodium-cooled fast reactors

Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki

Journal of Nuclear Science and Technology, 54(10), p.1036 - 1045, 2017/10

 Times Cited Count:2 Percentile:42.02(Nuclear Science & Technology)

To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium. The effect of use of the unstructured mesh was also investigated by the two analyses using structured and unstructured mesh.

Journal Articles

Experiments on gas entrainment phenomena due to free surface vortex induced by flow passing beside stagnation region

Ezure, Toshiki; Ito, Kei; Tanaka, Masaaki; Ohshima, Hiroyuki; Kameyama, Yuri*

Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 9 Pages, 2017/09

In the design of sodium cooled fast reactors, cover gas entrainment into sodium coolant (gas entrainment) is one of significant thermal hydraulic issues. This paper describes experimental results on surface vortex type gas entrainment which occurs in a share flow area where flow passes beside the stagnation region. In the experiment, the relationship between the free surface dimple shape and the velocity distribution around the free surface vortex was simultaneously grasped under several horizontal and suction velocity conditions by means of visualization measurement and Particle Image Velocimetry measurement. As the results, quantitative relationships among circulation, vertical velocity gradient and the gas core length were obtained in time-trends as fundamental data to develop the evaluation method of gas entrainment. Furthermore, it was confirmed that the evaluation method based on a vortex model, was an effective way to evaluate gas entrainment.

Journal Articles

Development of unstructured mesh-based numerical method for sodium-water reaction phenomenon

Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Watanabe, Akira*

Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 12 Pages, 2017/09

To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile and location of the Mach disk showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium.

Journal Articles

Thermal-hydraulics analysis of fuel assembly with inner duct structure of a sodium-cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki

Nippon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2017 Koen Rombunshu (CD-ROM), 4 Pages, 2017/08

A specific fuel assembly named FAIDUS (Fuel Assembly with Inner Duct Structure) has been developed as one of the measures to enhance safety of the reactor in the core disruptive accident (CDA) in JAEA. Thermal-hydraulics evaluations in FAIDUS under various operation conditions including the CDA are required to confirm its design feasibility. Therefore, numerical simulations by using thermal-hydraulics analysis program named SPIRAL developed in JAEA are conducted to analyze the thermal-hydraulics in the FAIDUS. Through the numerical simulation in the FAIDUS under tentative rated operation condition of an Advanced SFR, it was indicated that the flow and temperature distribution in the FAIDUS showed the same tendency as that in ordinary FA and specific characteristics was not observed.

JAEA Reports

Development of LEAP-III code for evaluation of long-time event progress under tube failure accident in steam generators

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki

JAEA-Research 2017-007, 61 Pages, 2017/07

JAEA-Research-2017-007.pdf:4.3MB

For safety assessment of a steam generator of sodium-cooled fast reactors, it is necessary to evaluate the possibility of occurring tube failure propagation and of water leak rate under sodium-water reaction accident. In the previous studies, a computer code called LEAP-II calculating a wastage-type failure propagation and the water leak rate during long-time event progress was developed. In this study, a numerical method to evaluate the possibility of occurring overheating rupture was introduced into the LEAP-II code to expand application range of this code. The completed code is called LEAP-III. The test analysis on a tube bundle configuration demonstrated that the overheating rupture model could provide conservative prediction.

Journal Articles

Thermal-hydraulic analysis of fuel assembly with inner duct structure of an advanced loop-type sodium-cooled fast reactor using ASFRE code

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07

In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.

Journal Articles

Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

Uwaba, Tomoyuki; Ohshima, Hiroyuki; Ito, Masahiro*

Nuclear Engineering and Design, 317, p.133 - 145, 2017/06

 Times Cited Count:1 Percentile:100(Nuclear Science & Technology)

The coupled numerical analysis of mechanical and thermal behaviors was performed for a wire-wrap fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal hydraulics analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that radial distribution of coolant temperatures in a subassembly tended to be flattened as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such temperature distribution was slightly analyzed as a result of the small bowing of the fuel pins due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal hydraulics was also investigated in this study.

Journal Articles

Identification of important phenomena under sodium fire accidents based on PIRT process

Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

The present PIRT process is aimed to identify key phenomena involved in sodium fire accidents that involve complex phenomena in sodium-cooled fast reactor plants. In this PIRT process, the figures of merit (FOMs) are specified through factor analysis. Associated phenomena are identified through the element- and sequence-based phenomena analyses. Importance of each associated phenomenon is evaluated by considering the sequence-based analysis of associated phenomena correlated with the FOMs. Then, we complete the ranking table through the factor and phenomenon analyses. An assessment matrix of important phenomena and experiments is completed finally for model validation.

Journal Articles

Progress of design and related researches of sodium-cooled fast reactor in Japan

Kamide, Hideki; Sakamoto, Yoshihiko; Kubo, Shigenobu; Oki, Shigeo; Ohshima, Hiroyuki; Kamiyama, Kenji

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures in order to strengthen safety of a fast reactor since the Great East Japan Earthquake. This paper describes the progress of design study and research and development related to safety enhancement and the severe accident countermeasures. For the purpose of strengthening of decay heat removal function, several researches have been carried out on the decay heat removal in a core disruptive accident (CDA), diversity and applicability of decay heat removal systems, and thermal hydraulic evaluation methods. In order to elucidate the behavior of molten fuel during CDA, some in-pile and out-of-pile tests has been performed by international collaboration including basic experiments. Core design was also improved from the viewpoint of preventing the occurrence of severe accident.

554 (Records 1-20 displayed on this page)